Small, fast neutron spectrum nuclear power plant with a long refueling interval

ABSTRACT

Nuclear reactor systems and methods are described having many unique features tailored to address the special conditions and needs of emerging markets. The fast neutron spectrum nuclear reactor system may include a reactor having a reactor tank. A reactor core may be located within the reactor tank. The reactor core may include a fuel column of metal or cermet fuel using liquid sodium as a heat transfer medium. A pump may circulate the liquid sodium through a heat exchanger. The system may include a balance of plant with no nuclear safety function. The reactor may be modular, and may produce approximately 100 MW e.

CROSS-REFERENCE TO RELATED APPLICATIONS

This application is a continuation of U.S. patent application Ser. No.15/583,838, filed May 1, 2017, which is a continuation of U.S. patentapplication Ser. No. 14/291,890, filed May 30, 2014, granted as U.S.Pat. No. 9,640,283, which is a continuation of U.S. patent applicationSer. No. 13/030,740, filed Feb. 18, 2011, granted as U.S. Pat. No.8,767,902, which claims priority to U.S. Provisional Patent ApplicationNo. 61/306,754, filed Feb. 22, 2010; the content of which areincorporated by reference herein in their entirety.

This application incorporates by reference in its entirety U.S. patentapplication Ser. No. 12/696,851, filed Jan. 29, 2010, granted as U.S.Pat. No. 8,571,167; the content of which is incorporated by referenceherein in its entirety.

STATEMENT REGARDING FEDERALLY SPONSORED RESEARCH OR DEVELOPMENT

The Government has certain rights in the invention pursuant to Work forOthers Agreement No. 854V0.

FIELD OF THE INVENTION

The present invention relates to nuclear power plants, and, moreparticularly, to fast neutron spectrum, sodium cooled reactors withmetallic fuel.

BACKGROUND OF INVENTION

World electricity demand is expected to as much as double by 2030 andquadruple by 2050. The world electricity demand increase is forecastedto come from developed countries but to an even larger extent fromdeveloping countries. To meet rapid growth in developing countries,nuclear energy should be packaged in a configuration tailored to meettheir specific needs.

BRIEF DESCRIPTION OF THE DRAWINGS

The accompanying drawings, which are included to provide a furtherunderstanding of the invention and are incorporated in and constitute apart of this specification, illustrate preferred embodiments of theinvention and together with the detailed description serve to explainthe principles of the invention. In the drawings:

FIG. 1 shows an exemplary Small Modular Reactor (“SMR”) according to anembodiment of the present invention.

FIG. 2 is an example of an SMR nuclear power plant according to anembodiment of the present invention.

FIG. 3 shows an exemplary nuclear energy architecture according to anembodiment of the present invention.

FIG. 4 shows an exemplary refueling cluster layout and core radialenrichment zoning according to an embodiment of the present invention.

FIGS. 5A-5B show an exemplary wedge used for core clamping andunclamping during refueling operations.

FIGS. 6A-6C show an exemplary wedge used for enhancing core radialexpansion reactivity feedback.

DETAILED DESCRIPTION OF THE EMBODIMENTS

A fast neutron spectrum, sodium cooled reactor with metallic fuel isdescribed.

FIG. 1 illustrates an exemplary Small Modular Reactor (“SMR”) system 501of the present invention. The SMR system may include a uranium-fueledcore 503. The core composition may be enriched (<20%) uranium/zirconiumalloy for the initial core and recycled uranium/transuranic zirconiumfor subsequent cores. Uranium 235/thorium/zirconium alloys may also beused in some embodiments.

The core 503 may be submerged in a tank 505 of ambient pressure liquidsodium 507. The tank 505 may be thin-walled stainless steel, and may besized for shipment by barge or rail. The tank 505 may be positioned in aguard vessel 517 and a deck 521 of the tank 505 that may be enclosed bya removable dome 519. The guard vessel 517 and dome 521 together maycreate a containment 523.

The SMR system 501 may be encased in a concrete silo 515. The core 503and its containment 523 may be emplaced in a concrete silo with aconcrete cover. The silo and its cover may create a shield structure toprotect the reactor system 501 and the containment 523 from externalhazards. The shield structure and/or the containment 523 and reactor 503may be seismically isolated.

The SMR system 501 may also include control rods 513.

The liquid sodium 507 from the tank 505 may be pumped by one or morepumps 509 through the core 503 to carry heat away from the core 503. Theliquid sodium 507 may carry the heat to one or more sodium to sodiumheat exchangers 511. The liquid sodium 507 may be heated from about 350°Celsius to about 510° Celsius.

FIG. 2 shows the SMR system 501 within a larger energy generation system601. The heated sodium 507 may pass through the heat exchanger 511 toheat secondary sodium, which in turn passes through a secondary heatexchanger 603 where the secondary sodium heats supercritical (almostliquid) carbon dioxide. The supercritical CO₂ is compressed to 21 MPa,just above its critical point at approximately 7 MPa and approximately31° C. It is then recuperated to ˜350° C. in regenerative heatexchangers 609; then further heated to ˜500° C. in the Na-to-CO₂ heatexchanger. The recuperation and compression of a nearly-liquid fluidallow for an approximately 40% energy conversion at a relatively lowtemperature compared to ideal gas Brayton cycles. The heatedsupercritical carbon dioxide may then be used to spin a gas turbine 605to make electricity in an electrical generator 608 in a carbon dioxideBrayton cycle building 607. The turbine 605 and compressor 606 rotatingmachinery is very compact owing to the high density of the CO₂. “Printedcircuit” heat exchangers used for recuperations and for sodium tosupercritical carbon dioxide heat exchange 603 are of extremely highpower density. Altogether the supercritical CO₂ Brayton cycle is muchmore compact than comparable Rankine steam cycle energy converters. TheBrayton cycle may provide the SMR a thermal efficiency (heat energyconverted to electricity) of approximately 39% to approximately 41% ormore, an efficiency much higher than conventional light water reactor(“LWR”) steam driven turbines. Furthermore, in certain embodiments ofthe present invention waste heat can be used to meet lower-temperatureneeds, such as space heating, water desalination, industrial processheat, or can be dissipated through cooling towers.

Small sodium-cooled fast reactors may demonstrate important inherentsafety characteristics. These reactors may operate with simplified,fail-safe controls that may facilitate rapid licensing by regulatoryauthorities. For example, in response to an accident condition, such asloss of coolant flow, overcooling in the heat exchanger, control rodrunout or loss of ability to reject heat, embodiments of the reactor mayshut themselves down without human or safety-system intervention. Forinstance, as the reactor coolant heats up, the core structures maythermally expand causing increased neutron leakage from the core, inturn causing power levels to decrease in a self-correcting fashion.

SMR operation requirements may be significantly simpler thanconventional nuclear systems due to a characteristic that allows thereactor to innately follow load requirements brought upon by varyinglevels of electricity demand.

Metal alloy fuel is well demonstrated, both from performance andfabrication perspectives, and can straightforwardly meet long refuelingtime interval requirements. Additionally, a cermet fuel may be used,while the cermet fuel none-the-less retains metallic alloy fuelattributes.

The reactor core may have a long life, up to about 20 years or more. Thereactor may not have or require permanent onsite refueling equipment orfuel storage capability. Refueling may be done by an outside serviceprovider that brings refueling equipment with a new core, changes thecore out, and takes both used core and refueling equipment away whencompleted. Fuel handling and shipping can commence at a very short timeafter reactor shutdown owing to the derated specific power (kwt/kgfuel). One or more multi-assembly clusters in a reactor core may havederated specific power (kwt/kg fuel) for enabling long refuelingintervals while remaining in the existing fuels database. This may alsoenable refueling operations very shortly after reactor shutdown.Refueling operations may start within approximately two weeks of overallreactor shutdown, and may finish within approximately 1 month of overallreactor shutdown. The whole reactor core may be replaced at one time,about every 20 years. As such, the reactor system may have norequirement that the operator handle fuel. The overall unit may besealed, physically and with electronic monitors, so that any intrusionattempt is easily detected. The elimination of any need or the abilityto gain direct access to the fuel and use of smart monitoring systemsnot only reduces operator requirements, but also addresses proliferationconcerns. Additionally, the SMR is small enough to be located belowground, which enhances containment and protection from terroristactivities. Finally, embodiments of the system are small enough thatthey can be shipped by barge, rail, and truck and installed at the siteusing modular construction techniques: this ability to remotelymanufacture and obtain economies of serial production is a desirablebenefit.

When the fuel cartridges are returned to themanufacturer/designer/fabricator's facility, nearly all of the usednuclear material can be recycled and used as fuel in future cartridges,greatly reducing the volume and radio-toxicity of the final waste to bestored in a geologic repository. Unlike used fuel from conventionallight water reactors, material from SMR's need not be stored for tens ofthousands of years. Non-recyclable materials from SMR's require only afew hundred years of storage before the waste decays to levels ofradiation associated with the original uranium ore.

The reactor concept and its supporting fuel cycle infrastructure mayoffer a configuration of nuclear energy tailored to meet the needs ofemerging electricity markets in developing countries as well as imminentglobal need for carbon-free non-electric energy sources. Thisconfiguration of nuclear energy may rely on the huge energy density ofnuclear fuel (>10⁶ times that of fossil fuel) to enable a distributedfleet of small fast reactors of long (20 year) refueling interval,providing local energy services supported by a small number ofcentralized facilities handling fuel supply and waste management for theentire fleet. The reactors may be sized for local and/or small grids,and are standardized, modularized and pre-licensed for factoryfabrication and rapid site assembly. Correspondingly, the centralizedfuel cycle infrastructure may be sized for economy of scale to support alarge fleet of reactors in the region and may be operated underinternational safeguards oversight. The configuration is tailored tomeet the tenets of sustainable development.

FIG. 3 illustrates an exemplary nuclear energy infrastructure in itsmature stage. A regional center 701 may supply/ship reactor fuel and/oraccept spent fuel returns from sub-regions, such as countries 703.Various regional centers 701 may trade in fissile and fertile materialto level out regional surpluses and/or shortages.

Reactor Overview

Embodiments of the present invention may include an approximately 50MW_(e) (125 MW_(t)) to approximately 100 MW_(e) (260 MW_(t))sodium-cooled fast reactor operating on a long (approximately 15 toapproximately 20 year) whole core refueling interval. An initial fuelload may be enriched uranium (<20% enriched) in the form of metal alloyfuel slugs, sodium or helium bonded to ferritic-martinsitic cladding.The reactor may exhibit an internal breeding ratio near unity such thatits reactivity burnup swing is small and its core is fissileself-sufficient. A burnup swing of less than approximately 1% Δk/k mayfacilitate passive safety and passive load follow. Embodiments of thepresent invention may attain 80 MW_(t)d/kg or more fuel average burnup,and upon pyrometallurgical recycle at completion of its 20 year burncycle, depleted uranium makeup feedstock may be all that is required forthe reload core. Upon multiple recycles, the core composition maygradually shift to an equilibrium transuranic fuel composition, which isalso fissile self sufficient, and thus requiring only U238 makeup uponrecycle.

A forced circulation heat source reactor may deliver heat at ˜500° C.through a sodium intermediate loop that drives a supercritical CO₂(S-CO₂) Brayton Cycle power converter attaining ˜40% conversionefficiency and may be capable of incorporating bottoming cycles fordesalination, district heat, etc. Other embodiments might drive aRankine steam cycle. Embodiments of the present invention may employpassive decay heat removal; achieve passive safety response toAnticipated Transients Without Scram (ATWS); and employ passive loadfollow. The balance of plant may have no nuclear safety function.

The plant may be sized to permit factory fabrication of rail and bargeshippable modules for rapid assembly at the site. Embodiments of thepresent invention may have features targeted to meet infrastructure andinstitutional needs of rapidly growing cities in the developing world aswell as non-electric industrial and/or municipal niche applications inall nations.

Targeting Emerging Markets

Nuclear energy is a well-established industrial business that, over thepast 35 years, has attained 13,000 reactor years of operating experienceand 16% market share of world electricity supply. Nuclear energy isbeing deployed primarily in the form of large size (greater than orapproximately equal to 1200 MW_(e)) plants in industrialized nations.There are currently 436 reactors deployed in 30 countries. Future growthin nuclear deployments is projected to be as much as 66% or even 100%additional capacity by 2030. The majority of the growth is projected totake place in developing countries where institutional andinfrastructure conditions often differ from those that, in the past,favored large scale plants and a once through fuel cycle. Developingnations often have small, local grids of under a few tens of GW, whichare unable to accommodate a 1.2 to 1.5 GW_(e) sized plant. Embodimentsof the present invention operating at 100 MW_(e), are not onlycompatible with smaller grid size but additionally, the smaller capitaloutlay required for its installation is compatible with a developingcountry's necessity for sharing limited financing across multipledevelopment projects during the early decades of its rapid economicgrowth.

A twenty year refueling interval with fuel supply, recycle, and wastemanagement services outsourced to a regional center enable a nation toattain unprecedented energy security even absent a need to first emplacea complete indigenous fuel cycle/waste management infrastructure.Moreover, centralization of fuel cycle facilities for economy of scalein technical and institutional safeguards operations may facilitate aninternational nonproliferation regime even for widespread worldwidedeployment of nuclear-based energy supply.

The energy supply growth rate in industrialized countries is projectedto be slower than in developing countries. Nonetheless, new nuclearplants are needed for replacements of coal and nuclear plants as theyare decommissioned at end of life. The large capacity interconnectedgrids in industrialized nations are compatible with large power ratingplants. Niche markets, however, are expected to rapidly emerge in bothdeveloped and developing nations for non-electric and/or cogenerationapplications of carbon-emission-free nuclear energy. Among these marketsmay be water desalination, oil sands/oil shale recovery and upgrading,and coal or bio to liquids synthetic fuel production. Passive safetyposture precludes any safety function being assigned to the balance ofplant and along with the reactor's reduced source term favor sitingadjacent to industrial and municipal installations.

Features of the Fuel Cycle

First, the core power density (kw_(t)/liter) and fuel specific powerkw_(t)/kg fuel may be derated so as to achieve a 20 year refuelinginterval while remaining within the bounds of the established metallicalloy fuels experimental database. This may provide a client long termenergy security and a high level of reliable availability.

Second, the once in 20 year whole-core refueling may be conducted byfactory personnel who bring the refueling equipment and fresh fuel fromoffsite, conduct the refueling operations, and then return the used coreand the refueling equipment to the factory. This may provide the clienta way to attain energy security absent a prior need to emplaceindigenous facilities for enrichment, fuel fabrication, reprocessing,and waste repositories.

Third, the refueling operations may be done on the basis of a fuelhanding assembly that may include multiple sub-components. Variousnumbers of sub-components may be included and may or may not beclustered. As an example, see an exemplary core made of seven fuelassembly clusters 801 in FIG. 4. FIG. 4 shows an exemplary arrangementof core components. For example, an outer layer of shield assemblies 803may cover a layer of reflector 805, which may cover a layer of outercore 807. Middle core 809 of a lower enrichment may generally surroundinner core 811 of still lower enrichment with primary control 813 andsecondary control 815 assemblies placed within the core 801. As shown,the fuel, reflector, shield and control rod assemblies are grouped intoseven-assembly clusters to speed the rate of core refueling.

During operations, the seven-assembly cluster may be transferred after avery short cooling period following reactor shutdown so as to minimallyinterrupt energy supply availability. The short cooling period andseven-assembly cluster features may be possible due to the derated fuelspecific heat (kw_(t)/kg fuel).

Fourth, the first fuel loading may be enriched uranium (enrichment <20%)and the core may be fissile self-sufficient such that at the end of the20 year operation interval, the core contains as much bred-infissionable content as has been burned out. Upon pyrometallurgicalrecycle of the used core, only U238 feedstock and fresh cladding may berequired for refabrication of a replacement core.

Fifth, over multiple recycles, the composition of the core may graduallytransition from a U235-rich composition towards an equilibriumtransuranic-rich composition that is also fissile self sufficient. Thefuel cycle waste stream may exclusively include fission products, whichrequire only 200 to 300 years of sequestration before decaying tobackground levels of radioactivity, whereas all transuranics may bereturned to the reactor as fuel where they are converted to fissionproducts.

Sixth, after the first core loading, all subsequent cores may requireonly U238 as feedstock. This may extend the world's ore resourcepotential to nearly 100% productive use, and yielding at least amillennium of energy supply. Capability to use thorium-based metallicalloy fuel extends the world's resource base to multi millennia.

Seventh, the fuel fabrication technology may offer the option ofincorporating LWR used fuel crushed oxide particles onto a metallicalloy to form a cermet. This option, when combined with an added (oxidereduction) step in the pyrometallurgical recycle process may offer aroute to cost effective management of LWR used fuel by subsuming it intothe fast reactor closed fuel cycle.

Features of a Heat Source Reactor

First, a core layout may include assembly clusters of individuallyducted and orifaced fuel assemblies. As described above, see FIG. 4 forexemplary seven-assembly clusters in a core layout. In otherembodiments, other numbers and arrangements may be contemplated. Theassemblies may be grouped into clusters for fuel handling whilepreserving individual fuel assemblies so as to retain the orificing andthe limited free bow reactivity feedback characteristics. Replaceablereflector and shield assemblies may be grouped into 3 or 4 assemblyclusters.

Second, a “limited free bow” core clamping approach may be used. Theclamping approach may utilize a removable and vertically adjustablehorizontal wedge 901 located in a central assembly position of a corelayout of ducted fuel assemblies 913 at an elevation approximately atabove-core load pads 903, as shown in FIG. 5A. Note that radialdisplacement as shown in FIGS. 5A and 5B is exaggerated. The wedge 901may be attached to a driveline 905 coupled to a vertical positioningmechanism 907 on a reactor deck 909. Preferably, the wedge 901 is at alower end of the driveline 905, where the driveline 905 is in a verticalorientation. The wedge 901 can be removed/withdrawn to loosen the corefor fuel handling, as shown in FIG. 5B. The wedge 901 can be re-insertedto clamp the core 915 and top load pads 917 outward against a coreformer ring 911 at a top load pad elevation once refueling is completed.Preferably the top load pads 917 may surround one or more ducted fuelassemblies 913 at approximately a top end of the ducted fuel assemblies913. The above-core load pads 903 may surround one or more ducted fuelassemblies 913 above a fuel elevation, but below the top load pads 917.A grid plate elevation may approximately correspond with a bottom end ofthe ducted fuel assemblies 913.

Third, a core may retain performance parameters, both operational andsafety, even as the fuel composition evolves over the 20 year burn cycleand further evolves from one recycle loading to another.

Fourth, embodiments of the present invention may include a strategy tomonitor reactivity feedbacks throughout core life and to fine-tune theirvalues using the vertical position adjustment of the wedge, should theydrift as the core ages over its 20 year burn cycle. The integralreactivity feedbacks may be measured in situ by non-intrusive smalladjustments of coolant flow rate, inlet coolant temperature, and controlrod position. The rest position of the core clamping wedge 901 may beused to adjust the value of a core radial expansion component of theinherently negative power coefficient of reactivity, as shown in FIGS.6A-6C. Note that radial displacement as shown in FIGS. 6A-6C isexaggerated. As shown in FIG. 6A, increasing power may increase outward(towards the right in FIGS. 6A-6C) bowing 951 of fuel assemblies 913.Unrestrained flowering upon an increase in core power may result from anincrease a radial thermal gradient on the ducted fuel assemblies 913.Inboard ducted fuel assemblies 913 may push outward, as shown in FIG.6B. Limited free bow core restraint may enhance radial dilation at fuelzone elevation of ducted fuel assemblies 913. As shown in FIG. 6C, anincrease in coolant outlet temperature may bathe the wedge driveline 905with increased temperature such that the driveline's thermal expansionmay drive the wedge 901 downward/deeper. This may in turn amplify theradially outward bowing of core fuel assemblies 913 at a fuel zoneelevation, which then may increase axial leakage and reduce reactivity.By adjusting a rest position of the wedge 901 at full power and fullflow, the amplitude of the bowing enhancement can be fine tuned.

Fifth, a passive safety response may be provided for loss of flow, lossof heat sink, chilled coolant inlet temperature and single rod runouttransient overpower (ATWS) transient initiators without scram. Theinnate reactivity feedbacks with respect to power and fuel and coolanttemperatures, when combined with a nearly zero reactivity burnup swingand with natural circulation capability at decay heat levels, may takethe reactor to an undamaged safe state for all ATWS initiators, i.e., nodamage may be incurred and a stable state may be reached for theseinitiators even if the rods fail to scram.

Sixth, a passive decay heat removal channel may be provided to theambient atmosphere ultimate heat sink always operating as a backup toactive decay heat removal channels. The passive channel may always beoperating at less than or approximately equal to 1% full power and canbe confirmed to be functioning at all stages of core life by in situnon-intrusive measurements. The heat capacity of the core and internalstructure is sufficient to safely absorb the initial transient of decayheat in excess of the passive channels' capacity.

Features of a Power Plant

First, a heat source reactor driving a S-CO₂ Brayton cycle energyconverter may attain nearly 40% or more heat to electricity conversionefficiency while operating in the working fluid range of ˜500° C., 21MPa to 31° C., ˜7 MPa. This converter may use rotating machinery ofextraordinarily high power density and recuperative heat exchangers ofexceptionally high power density.

Second, a heat source reactor may passively load follow the energyconverter demand for heat. The reactor may sense the balance of plantdemand communicated via flow rate and return temperature of theintermediate heat transport loop. The reactor's innate reactivityfeedbacks may maintain heat production in balance with heat removalthrough the intermediate loop within tens of seconds and without needfor active adjustments of control rods.

Third, a Balance of Plant (BOP) may be provided that carries no nuclearsafety function and can be built, operated and maintained to normalindustrial standards. The reactor can passively accommodate allphysically attainable combinations of flow rate and return temperaturereturning from the BOP through the intermediate heat transport loop. Thepassive decay heat removal channel may have no dependence on the BOP,and the nearly zero burnup control swing makes a rod runout TOPresulting from a control system error a no damage event. So the BOP neednot carry any nuclear safety function.

Fourth, embodiments of the present invention may include a potential totie a broad diversity of BOP configurations to a standard, pre-licensedheat source reactor since the BOP carries no nuclear safety function.The S-CO₂ Brayton cycle may reject ˜60% of supplied heat and may do sobetween ˜100° C. and 31° C. Many cogeneration options may exist for sucha temperature range, including multi-effect distillation desalinization;district heat; district chilled water; ice production and others.Alternately, diverse non-electric industrial processes may be co-sitedclosely with the heat source reactor, given its self-protectionfeatures, small source term, passive load following feature, and highlevel of availability.

Although the foregoing description is directed to the preferredembodiments of the invention, it is noted that other variations andmodifications will be apparent to those skilled in the art, and may bemade without departing from the spirit or scope of the invention.Moreover, features described in connection with one embodiment of theinvention may be used in conjunction with other embodiments, even if notexplicitly stated above.

1.-24. (canceled)
 25. A wedge for a core clamping system, the wedgecomprising: a first end and a second end wherein the first end has lesssurface area than the surface area of the second end; a drivelinecoupled to a vertical positioning mechanism connected to a reactor deck;wherein the wedge is vertically adjustable; and said wedge is configuredto loosen a nuclear fuel core assembly for fuel handling.
 26. The wedgeof claim 25, wherein the wedge is at a lower end of the drive line. 27.The wedge of claim 25, wherein the wedge is inserted into the nuclearfuel core assembly
 28. The wedge of claim 27, wherein the wedge iscapable of re-insertion to clamp the nuclear core fuel assembly.
 29. Thewedge of claim 25, wherein the wedge can be inserted to clamp thenuclear core fuel assembly and top load pads outward against a coreformer ring at a top pad elevation.
 30. The wedge of claim 29, whereinthe top load pads surround a ducted fuel assembly.
 31. The wedge ofclaim 30, wherein the top load pads surround the ducted fuel assemblyabove a fuel elevation and below the top load pads.
 32. The wedge ofclaim 25, wherein a vertical position of the wedge is adjustedthroughout the nuclear core fuel assembly's life to adjust the corereactivity values.
 33. The wedge of claim 25, wherein the verticalposition of the wedge is adjusted to loosen the nuclear core fuelassembly.
 34. The wedge of claim 25, wherein the vertical position ofthe wedge is used to adjust nuclear core fuel assembly radial expansion.35. The wedge of claim 25, wherein the vertical position of the wedgemay be in response to increased core coolant temperature.
 36. The wedgeof claim 35, wherein the wedge's vertical position is moved downwardfrom the wedge's initial vertical position in response to the increasedcore coolant temperature.
 37. The wedge of claim 36, wherein thedownward position of the wedge causes increased axial leakage andreduced reactivity of the nuclear core fuel assembly.